ML18045A677
| ML18045A677 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 09/26/1980 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18045A678 | List: |
| References | |
| IEB-79-14, NUDOCS 8010020367 | |
| Download: ML18045A677 (23) | |
Text
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consumers Rower company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201 * (517) 788-0550 September 26, 1980 Direct6r, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 US Nti~lear Regulatory Commission Washington, DC 20555
- DOCKET 50-255 - LICENSE DPR PALISADES PLANT - FINAL RESPONSE TO IE BULLETIN 79-14 IE Bulletin 79-14, Seismic Analysis for As-Built Safety-Related PipingSystems, requested that we verify that the seismic analysis input information conforms to the actual configuration of safety-related systems for our Palisades Plant.
We have provided interim information on this bulletin as shown in Attachment 1.
The "Final Response to NRC IE Bulletin 79-14 for Consumers Power Company Palisades Plant" is being transmitt~d to you as Attachment 2 of this letter.
As a result of the work. done for this bulletin, certain pages of the Palisades FSAR were revised.
These revised page ch.anges will be transmitted in a separate letter and are also included here as Attachment 3 (in DRAFT form) for your information.
The revised drawings resulting from the work done on the subJect bulletin will accompany the signature letter only.
David P Nuclear Licensing Administrator CC Director, Region III, USNRC
. - NRG Resident lnspector-PaHsades--
Attachments*
8010020
I
- 1.
IE Bulletin 79-14 dated July 2, 1979
- 2.
CP Co letter to NRC dated August 1, 1979
- 3.
CP Co letter to NRC dated August 31, 1979
- 4.
CP Co letter to NRC dated November 9, 1979
- s.
CP Co letter to NRC dated January 25, 1980
- 6.
CP Co letter to NRC dated February 14, 1980
- 7.
CP Co letter to NRC dated February 27, 1980
- 8.
CP Co letter to NRC dated March 11, 1980
- 9.
CP Co letter to NRC dated April 14, 1980
FINAL RESPONSE TO NRC IE BULLETIN 79-14 *
. FOR CONSUMERS POWER COMPANY
. PALISADES NUCLEAR PLANT
- Revision 0 July 22, 1980
- 1.
INTRODUCTION
- 2.
ORGANIZATION 3.. SYSTEM IDENTIFICATION.*
- 4.
INSPECTION A.
PERIOD OF INSPECTION B.
ELEMENTS OF INSPECTION C.
RESULTS OF INSPECTION D.
INSULATION REMOVAL
- 1)
Sampling Program
- 2)
Boric Acid System
- 3)
Modifications E~
ACCESSIBILITY CONTENTS
- 1)
Reactor Coolant. System
- 2)
Containnient Spray
.. 3)
- Other
. F... VERIFI.CATION OF PIPE WALL THICKNESS AND
. PIPE WELDS
- 5.
ANALYSIS A.
METHODS OF ANALYSIS
.* e Palisades 79-14 Rev O 7/22/80 1
1 1
3 6
- 1)
Comparison of Data with Existing Analysis
- 2)
Reanalysis.of Large Pipe
- 3) Reanalysis of Piping Supports
... *B-.- *-** REACTOR-*cooLANT -sYSTEM-ANALYSis*:
C.
RESULTS OF ANALYSIS
- 6.
MODIFICATIONS 10
- 7.
SMALL PIPING 10 a~* RESOLUTION OF INSPECTION ITEMS 11 i
9.*. - LICENSEE* EVENT REPORTS (LERs)
- 10.
CONCLUSIONS APPENDIX A
Drawings Defining Safety Piping Systems (Provided With Original - Signature Copy - Only)
Drawing Listing Palisades 79-14 Rev 0 7/22/80 12 12 13 14
- --~---* *---
-:-,+-
ii
SEISMIC ANALYSIS FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS
- 1.
INTRODUCTION Palisades 79-14 Rev 0 7/22/80 U.S. Nuclear RegUlatory Commission (NRC) IE Bulletin 79-14, dated July 2, 1979, requested information related to safety-related piping systems.
The bulletin was revised and supplemented by the following:
a)
Revision 1, issued on July 18, 1979 b)
Supplement 1, issued on AugUst 15, 1979 c)
Supplement 2, issued on September 7, 1979 The bulletin requested responses at 30, 60, and 120 days after July 2, 1979.
These ~esponses have previously been supplied. to the NRC.
An inspection of approximately 18,100 feet of safety-related
- piping, approximately 1,550 pipe supports, and piping components (2-1/2 inches and larger) at the Palisades Nuclear Plant has
- been conducted, and the items noted during the inspection have been dispositioned.
Additionally, reanalysis has beeri performed on the systems.
Small piping systems (2 inches and smaller) were excluded by the bulletin.
However, these systems were also inspected, noted.
items evaluated, and a sample of small piping systems was evaluated.
The following report describes the significant elements of the program and the results.
- 2.
ORGANIZATION Consumers Power Company (CPCo) contracted with Bechtel
- . Associates Professional Corporation, Ann Arbor, Michigan, to perform the inspection-and evaluation.
- 3.
SYSTEM IDENTIFICATION Current Palisades systems were reviewed, and those systems having safety-related piping are listed below, along with the applicable piping and instrumentation diagram (P&ID) drawing numbers.
1
P&ID Number M-201 M-202 M-203 and M-204 M-205 M-207 M-208 M-209 M-210 M-211 M-212 M-213 M-214
- M-215 M-218.
M-219 M-220 M-221 M-222 M-224 M-225 M-226 M-650 M-651
-~. Palisades i9-14 Rev o 7/22/80
System Description
Primary coolant system Chemical and volume control Safety injection, containment spray shutdown cooling Main steam and auxiliary turbine system Feedwater and condensate system Service water system.
Component cooling water system Radwaste treatment system - *clean Radwaste treatment system - dirty and
- gaseous Service and instrument air Circulating water, screen structure chl6rination, and fire system Lube oil, fuel oil, and diesel generator Plant heating system Heating, ventilatirtg, and air-conditioning system.
Sampling system Makeup water, domestic water, and chemical injection Spent fuel pool coolant and shield cooling
- system Miscellaneous gas supply system Gas. ~n.ci~yzing. s:y~t:em High-pressur~, air-operated valves Steam generator blowdown modification Radwaste evaporator system - clean.wastes Radwaste evaporator system - miscellaneous wastes 2
Palisades 79-14 Rev O 7/22/80 Appendix A of this report provides reduced copies of the record
- -drawings which delineate the boundaries of Seismic Category I piping on the various system P&IDs.
- 4.
INSPECTION A.
PERIOD OF INSPECTION Inspection was performed at the Palisades plant during the period of July 15, 1979, to May 1, 1980.. As-built piping isometrics and support sketches were originated at the site.
Finalization of these drawings is being completed at Bechtel's office in Ann Arbor, Michigan.
B.
ELEMENTS OF INSPECTION
- c.
- 1)
The inspection was performed according to a field procedure which detailed the various aspects of.
inspection.
- 2)
The essential items verified by the inspection are*as follows:
a)
Pipe geometry b)
Support design, function, location, and clearance c)
Pipe attachments d)
Valve and valve operator location and orientation RESULTS OF INSPECTION In* accordance with the field procedure,.data and* sketches
. showing the existing safety-related piping sections were.
compl~ted. **Potential nonconformance items were also listed..
These items are further discussed in Section 8.
The as-0 built data were evaluated as discussed in Section 5.
D.
INSULATION REMOVAL
- 1)
Section 2.A.l of CPCo's 30-day response to this bulletin noted that thernial insulation would not be removed to facilitate inspection.
The basis.for this response is that Palisades has an ongoing inservice inspection (ISI) program which includes periodic removal of insulation _.
and lrispection* of a "signific.ant ntunber ""of-p:lpe supports and attachments welded to piping.
Therefore, specific removal of insulation to meet the. intent of NRC IE Bulleti.n 79-14 was not deemed necessary.
However,. the following items were, in fact, accomplished during the inspection.
3
E.
Palisades 79-14 Rev o 7/22/80 a)
A sample of 20 supports which were partially insulated were selected, based on the following
- criteria..
o supports were from at least four separate functional (as opposed to piping). systems.
o Approximately half of the supports were pipe clamps covered by insulation; the other half were supports with attachments welded to pipe covered by insulation.
- o Consideration was given to ALARA when selecting the supports.
Insulation was removed at the 20 support.locations, and the support configuration compared with the*
drawing.
b)
Installation of new pipe heating elements during the outage resulted in removal of insulation on the heat-traced portion of the chemical and volume control system.
Approximately 51.additional supports in this area were compared with the drawings.
- c)
As part of the normal !SI program, insulation was removed from numerous other hangers, allowing inspection of hanger geometry.
d)
A number of supports were modified (see Section 6),
which required insulation removal.
The existing.:
supports were compared with the drawings.
- 2)
No significant items were.found as a result of the*
insulation removal program.
- 3)
Based on.the above discussion, adequate confidence was established that further removal of insulation would not reveal significant inspection anomalies..
ACCESSIBILITY The total length of large safety-related piping (2-1/2 inches and larger) at the Palisades plant is approximately 19, 150 feet.
Approxi~a_t~ly 95 percent _o{.the**
T9, 150 feet was iri.s-pected in *accordaw::e with the field procedure.
The remainder of the pipe (approximately 1~050 feet) is discussed as follow~. *
- 1)
Reactor* coolant system piping from the outer surface of the biological shield wall to the reactor vessel was
. deemed inaccessible because of high radiation fields.
The inaccessible reactor-coolant system pipe totaled 4
' 2.)
Palisades 79-14 Rev b 7/22/80 approximately 50 feet~ The measurements taken and their relationship-to analysis are discussed in Section 5.B.
The containment spray system piping consists mainly of vertical headers on the containment inner wall, horizontal headers near the top of the containment, and ring headers off the horizontal runs.
The.majority of both the ring and horizontal headers was deemed inaccessible because of the excessive scaffolding which would have been required to closely inspect the piping and components.
The containment spray system's inaccessible piping totaled approximately 975 feet.
. The following* items were completed with respect to the containment spray.piping.
a)
The vertical headers and the two ring headers near the containment inner wall were inspected to the field procedure.
b)
The remainder of the piping was compared with the drawings from the nearest vantage points.
- The
- piping systems are generally symmetric; therefore, visual comparison of estimated distances and comparison with struts and beams provided adequate confidence in the piping geometry.
c)
Photographs were made (usl.ng telescopic lenses) of each pipe support.
The photographs were used to verify the geometry given by as-built pipe support shetches for the containment spray system.
- 3)
The remainder of the piping deemed inaccessible included.*
various short runs of pipe.
This piping.totaled approximately 30 feet.
This piping was also viewed, as appropriate, from the nearest vantage points to gain confidence that the recorded geometry does not have.*
significant deviations.
- 4)
Based on the above, a high confidence level.has been achieved regarding that portion of piping deemed inaccessible.. No further inspection is considered necessary.
F.
VERIFICATION OF PIPE WALL THICKNESS AND PIPE WELDS
-- - - --- - 1 r--The-Pal1sadei:f-pianf---has an origoirig I SY -pr'ogram---which -
periodically examines selected piping welds and adjacent material ultrasonically.. Approximately 90 percent of all !SI ultrasonic data has been reviewed.
The followj,.ng points.are noted.
a)
Wall thickness readings are taken upstream and downstream of an inspected weld and recorded on a lamination scan data sheet.
5
- 5.
- 9 Palisades 79-14 Rev O 7/22/80 b)
Fabrication tolerances, referenced in the Pipe Fabrication Institute Technical Bulletin TBl-1974, are understood to be 12-1/2 percent of the nominal wall thickness.* The inservice inspection identifies the weld, type of weld (e.g., pipe-to-pipe weld or pipe-to-elbow weld), the nominal wall thickness.
specified, *and the wall thickness measured.
The difference between these last two items is expressed as a percentage.
c)
The review has indicated that wall thicknesses measured at the Palisades plant are within manufacturing tolerances and conform to the schedule of piping as specified.
In some cases the piping was found to be; thicker, representing conservative installation prac:tices.
In most.cases the thicker piping was found at expected locations; e.g.,
elbows, tees, and valves).
No data points were found to be thinner than the 12-1/2 percent
.tolerance allows.
d)
The data reviewed represent measurement of 78 welds and 156 adjacent pipe locations.
- 2)
Adequate confidence was achieved based on the. above such that no additional inspection of piping.wall thicknesses or welds was deemed appropriate.
Specified wall thicknesses used in piping stress.reanalysis are in compliance 'With actual as~built conditions.,*
ANALYSIS U.S. NRC IE Bulletin 19-14 required verification that seismic piping analysis input information conformed to.the actual*
configuration of safety-related systems..
- Prior. to collecting the as-built data at the Palisades plant, the.,existing.analysis was reviewed for completeness.
The*analysis completed for the original* plant construction.was completed by category; i.e., one group completed.the deadweight and pipe support analysis, another*did*thermal, and the.third group completed the seismic analysis.
The deadweight and pipe support analysis.was not retrievable~ Much of the thermal and.seismic.analysis was available; however; such items as the.use of different piping system boundaries and the use. of different node. points Iriade correlation very difficult.
Therefore,* the decision was made
- and implemented to reanalyze the as-built data (piping and
-supports-) -a-s-;0 they were* acctlriiulate:d-. -- The-refanalysis-exceeds- -the-- -- -**--
requirements of U ~ s. NRC IE.* Bulletin 79-14; however, it was performed to provide CPCo with a consistent analytical base for.
use in conjunction with the ongoing Systematic Evaluation Program (SEP) and for use during future plant piping
- -mo(j.ifications.
The. reanalysis is discussed below.
6
Palisades 79-14 Rev O 7/22/80 A.
REANALYSIS (PIPING 2-1/2 INCHES IN DIAMETER AND LARGER)
B.
A total of 71.piping stress systems- (i.e., portions of total functional systems routed between fixed anchor or terminal connection points) have been identified as applicable within the scope of this program.
The reactor coolant system (piping between the reactor vessel, the steam generators, and the pressurizer less the pressurizer surge and spray lines) is unique and is discussed in Section 5.B.
The following discussion pertains to the remaining 70 piping stress systems.
The Palisades Final Safety ¥alysis Report (FSAR) defines U.S. ANSI B31.1. O ( 1967) as the applicable code for original plant design and analysis.
Because that code.did not define*
stress limits for the faulted condition (SSE, 0.2 g earthquake) and because no recognized criteria for the
- faulted condition existed, Appendix A to the Palisades FSAR (Section A.2.b) established 1.1 Sy as the allowable stress for the faulted condition.
The FSAR criterion was very consenrative as evidenced by later code requirements.
A better understanding of materials and computer techniques,
- together with increased computer capacity, resulted in subsequent code editions with reduced stress intensification factors (see ASME Code, Sect;.ion I I I, NC 3652,* Equations 8,
- 9, and ll), and 2.4 Sh being established as a limit for the faulted condition (SSE, 0.2 g earthquake).
Because the 2.4 Sh factor *and the modified stress intensification factors were included in later codes and have been accepted for use on later nuclear plant analysis,.the factors are considered acceptable for the Palisades plant reanalysis and for future analysis during the remainder of the plant's life..
Based on the above, piping stress packages will be completed to either of the following criteria:
- 1)
Criteria as defined by the FSAR (i.e., USAS B31.1. O, 1967 edition plus Appendix A of the FSAR)
- 2)
Criteria as defined by the FSAR, except that 2.4 sh is used for.the faulted condition and stress intensification factors in accordance with the ASME Code,Section III, NC 3652 (Equations 8, 9, and 11) in lieu of USAS B31.l.O, Section 119~6.4 are used.
Work on the pipe stress calculations 1ising the as-built configuration inspection data has been in progress and will be completed in September 1980.
REANAL~SIS OF PIPE SUPPORTS Reanalysis of approximately 1,550 pipe supports is being completed using the as-built sketches and data, together with loadings which resulted from the piping reanalysis 7
'9 Palisades 79-14 Rev O 7/22/80 discussed above.
Two categories of the analysis are as follows:
.1)
The supports were.analyzed and calculations completed for each support.
- 2)
- The supports were evaluated on a judgment basis by qualified, trained engineers.
If the particular support was judged adequate, no further analysis was performed.
If the support was judged inadequate, then a calculation was completed on the support.
Calculations were completed for new supports required by piping stress analysis.
C.
EVALUATION OF REACTOR COOLANT SYSTEM A review of the main loop piping analysis (Analysis Report for CPCo Piping CENC 1115) indicated that the reactor vendor checked the geometric accuracy of the actual piping layout and showed that the vendor checked flexibility analysis input against the layout drawings.
Further evaluation revealed that the main loop piping to* be designed and analyzed for seismic loadings on.the basis of equivalent "g" loads applied to flexibility analysis models.
Input to the program consisted of system geometry (piping components and supports), material properties, and seismic loadings.-
Review of individual piping sections defined by Drawing E-232-673, and the assembled piping system. defined by Drawings 2966-SW-2066 and 2966-SE-2067; shows that the main loop piping is made. up of short, large-diameter, and basically simple runs of piping having no intermediate supports or hangers.
- Support for the main loop.. piping is provided by* the interconnected major components.* Because of*
the short, essentially simple nature of the main loop piping, the setup of the piping pieces determines the final as-built locations of the major components.
Conversely, the proper operation of tpe component supports which are designed to allow unrestrained thermal expansion of the ma~n loop piping.ensures that the piping is installed to within close. tolerance limits of the theoretical locations.
Proper function of the major component support was established and confirmed during hot functional testing.
-Even though -ehe *-evaluation/review -effort-above* sfo)ws
- adequate confidence was established that the as-built primary coolant system meets the design* requirements, it was decided that a portion of the main loop piping be.inspected to ensure that the input to the seismic analysis is correct
~*an'd will meet the requirements* and intent of NRC IE Bulletin 79-14.
8
Palisades 79-14 Rev O 7/22/80 Using the drawings referenced above, inspection personnel verified the d1mensional and geometric accuracy as it relates to the actual configuration of the main loop piping in containment, specifically in the low radiation areas outside and up to the shield wall encompassing the reactor vessel.
Further inspection showed that*the major as-built component locations were as defined by the drawings.
- Thus, by veri~ying that the major as-built locations agree with the referenced drawings, we judge that the input to the seismic analysis is confirmed.
From the above discussion, it is concluded that the inspection of the main loop piping inside the shield wall encompassing the reactor vessel is not necessary and that ALARA criteria have been met by eliminating high radiation exposure to inspection personnel that would have been required for that area.
D.
RESULTS OF REANALYSIS
- The results of the reanalysis of the 70 stress packages and related supports showed that a significant number of piping systems did not meet the criteria as discussed above.
Approximately 320 support changes have been made; an estimated 100 more support changes will be incorporated, 10 piping junctures were reinforced, and several containment dome truss modifications were completed.
These modifications are discussed in more detail in Section 6.
With respect to piping stress analysis, some of the causes*
of deficiencies are attributed to the generic limitations (in scope) of seismic computer programs in use at the time of.the original design of. this project.
The computer
- program limitations.resulted in analyses which included decoupling, overlapping, and nonconservative assumptions in mass point selections for portions of.some systems.
In addition, some of the systems.were analyzed using an insufficient number of dynamic modes.
Under this program, the combined stress system calculations showed the early calculation approaches to be less conservative. Also, a significant portion of the deficiencies concerned the.
containment spray system.
The original plant.seismic analysis for the containment spray system was a simplified hand calculation which did not account for such items as interaction and stress intensification.
With respect to*pipe.support analysis, the earlier calculations are not available; therefore, a comparison
.cannot be made.
E.
INTER.IM OPERATION An assessment of the analysis and modifications was made in early 1980.
CPCo proposed and the NRC concurred with an interim.operation of the Palisades plant after completion of 9
Palisades 7.9-14 Rev O 7/22/80 certain analytical and modification efforts.
These are
docl1mented as follows:
.1)
CPCo letter to the NRC, dated February 14, 1980, D.P. Hoffman (CPCo) to D.C. Zeimann (NRC), with attachments
- 2)
NRC letter to CPCo, dated April 25, 1980, D.C. Zeiman (NRC) to D.P. Hoffmann (CPCo), with enclosure 6;
MODIFICATIONS A.
As a result of the analysis discussed in Section 5, a significant number of piping and support modifications were made.
B.
Approximately 320 piping support changes have been completed to date and are summarized as followf:;:
- 1)
A total of 45 new supports were added.
- 2)
A total of 23 supports were removed.
- 3)
A total of 252 supports were modified.
C.
In addition to pipe support changes, 10 piping junctures were reinforced either by ad.di ti on of reinforcment saddles or by addition of weld at the juncture.
D.
Final*ly, reanalysis of the containment spray system supports and their connections with trusses in the containment dome region resulted in modifications as follows (Reference Drawing C-38, Revisions 7 and 8, containment liner support trusses):
- 1)
Reinforcement of the diagonal at the containment wall end of trusses at azimuths 175 and 355 degrees
- 2)
Reinforcement of the bottom chord and a W-8 beam at azimuth 15 degrees E.
The reanalysis process in accordance*. with Section 5 has not been completed~ The current estimate is that approximately 100 more supports wi.11 be. made as a result of the remaining reanalysis.
- 7.
-SMALL PIPING U.S. NRC IE Bulletin 79-14 excludes small piping (2 inches in diameter and less) from its scope.
The small piping at Palisades was, however, also inspected, and as-built drawings were completed.
Documents on the original criteria for small piping were reviewed, and a sample of 11 as-built drawings were compared with the criteria.
No significant deviations were 10
Palisades 79 Rev O 7/22/80 found.
The as-built isometric drawing numbers of the 11 samples
-*.are listed as follows:
Isometric Drawing SP-FSK-GC9-2 M599-1, Sh 27 SP-FSK-HC27-3 SP-FSK-CC13 SP-FSK-CC13-4 SP-FSK-CC4-3 SP-FSK-HCl-4 SP-FSK-CCl-1 SP-FSK-CC3-1 SP-FSK-EB-11-8 SP-FSK-HC24-88 Description Safety Injection Fill and Drain High Pressure Safety Injection CCW Seal Leakof f from RC Pump P50-B to Primary Drain Tank T-74 Control Bleedof f from RC Pump P50-C Control Bleedof f from RC Pump P50-A SI Tank T82C to Pressure Transmitter Reactor Inner Leak Test *Line From CV-0101 to Drain Tank T-74 Collection Hdr.
Reactor Inner Leak Test Line From Biological Shield to CV-0101 Letdown Line from Regen Heat Exch.
E-56 to Letdown Flow Control Valves Surface Blowdown from S/G E-SOB -
from Cont. Pen. #55-to Blowdown Control
. Valve CV-0738 Quench Tank Drain to Drain Tank T74 Collection Hdr.
- 8.
RESOLUTION OF INSPECTION ITEMS The.field walkdown procedure includes a listing of questionable or discrepant conditions.
These items (approximately 3,250 for large and srnail bore pipe) were.evaluated and resolved. - The categories of items are discussed as follows:
A.
Approximately 2,500 of the 3,250 total items related to lack
- of, or nonconforrnance with, existing drawings for such i:terns as piping, supports, and penetrations.
These i terns were*
resolved by issuing current drawings which were used in the reanalysis discussed in Section 5 above.
B.
The remaining 750-iterns related to existing hardware conditions.
Examples of the conditions are as follows:
- 1) - Nuts and/or bolts - loose, missing, or without full engagement
- 2)
- 3)
Spring cans bottomed out or with no }oad
'.Bent, broken, or missing pipe support*cornponents These 750 maintenance type findings were evaluated and appropriate resolutions made; i.e., missing nuts and/or bolts were replaced, spring cans were adjusted, and missing components were replaced.
Resolution, of the inspection items ensured that the piping
- and supports reanalyzed in accordance with Section-5 above 11
Palisades 79-14 Rev O 7/22/80 matched with the analytical assumptions that the supports
_ were as depicted by the drawings and were functional.
- 9.
LICENSEE EVENT REPORTS (LERs)
Two LERs were issued as a result of the program implemented to meet the intent of U.S. NRC IE Bulletin 79-14.
A.
LER 79-033 described a deformed support to which a snubber was attached.
Corrective action has been completed.
B.
LER 80-001 noted that several piping systems did not meet
- FSAR stress criteria.
Corrective action on this LER is considered complete with submittal of this report.
10.* CONCLUSIONS Based on the discussions in the preceding sections of this report, the following conclusions are noted.
A.
The intent of U.S. NRC IE Bulletin 79-14 was to ensure that existing analysis of large piping reflected the as-built or installed condition.
CPCo has met and exceeded the intent of the bulletin for the Palisades plant by generating current as-built drawings and performing reanalysis using the as-built data.
B.
The criteria for reanalysis discussed in Sectioh 5 is appropriate for the Palisades plant.
- c.
Appropriate corrective actions have been taken for interim operation until any further required changes are implemented (see NRC letter, dated April 25,.1980, D.C. Zeimann (NRC) to.
D.P. Hoffman (CPCo) and attached safety assessment).
Any further required changes will be completed prior to the end of the next refueling outage.
D.
When work described in this. report is complete, all anomalies resulting from the U.S. NRC IE Bulletin 79:-14 effort will be considered complete.
12
Drawing No MEXP-201 MEXP-202 MEXP-204 MEXP-204 MEXP,..205 MEXP-207 MEXP-208 MEXP-209 MEXP-210 MEXP-211 MEXP-212 MEXP-213 MEXP-214 MEXP-215 MEXP-218.
.. MEXP-219 MEXP-220 MEXP-221 MEXP-222 MEXP-224 MEXP~225 MEXF-226 MEXP-650 MEXP.;.651 DRAWINGS DEFINING SAFETY PIPING SYSTEMS
- Revision No 3
4 1
2 1
1 1
2 3
1 1
2 2
1 1
2 2
3 1
4 2
- 1 1
1 Piping and Instrument Diagram Blueprints
- Primary Coolant System Chemical &*Volume Control System Safety Injection, Containment Spray & Shutdown Cooling System Safety Injection, Containment Spray & Shutdown Cooling System Main Steam, Main & Auxiliary Turbine Systems Feedwater & Condensate System~
- Piping & Instrument Diagram Service Water System Component Cooling System Radioactive Waste Treatment System Radioactive Waste Treatment System - Dirty
.Waste & Gaseous Waste Service & Instrument Air Systems Circulating Water, Screen Structure, Chlorination & Fire Systems Lube Oil, Fuel Oil & Diesel Generato~ Systems Plant Heating System Heating, Ventilation & Air Conditioning Process Sampling System Makeup Domestic Water & Chemical Injection Systems Spent Fuel Pool Cooling & Shield Cooling Systems Miscellaneous Gas Supply Systems Gas Analyzing Systems High-Pressure Aii Op~rated Valves Steam Generator Blowdown Modification Radwaste Evaporator System Clean Wastes
- Radwaste Evaporator System Miscellaneous Waste 14
I*.
DRAFT Revised Pages to the Palisades Plant FSAR Resulting From IE Bulletin 79-14
~Palisades !EB 79-14
Palisades !EB 79-14 The final design of the missile barrier and equipment support structures inside the containment is reviewed to assure that they can withstand applicable pressure loads, jet forces, pipe reactions and earthquake loads without loss of function.
The deflections or deformations of structures and supports are checked to assure that the functions of the containment and engineered safeguards equipment are not impaired.
Missile barriers are designed on the basis of absorbing energy by plastic yielding.
(b)
Class I Systems and Equipment Design (1)
Class I systems and equipment including main reactor coolant p1p1ng (excluding reactor vessel internals under DBA loadings and other Class !_piping and piping supports) are designed to the following criteria for load combinations and stresses:
Loads Stresses
- 1.
MOL Plus. PIT Plus SL Applicable Code Allowable Stress
- 2.
MOL Plus MIT. Plus SL Minimum Yield Stress at Temperature *
- 3.
MOL Plus MIT Plus 2SL Minimum Yield Stress at Temperature May Be Exceeded but Limited to No More Than Plus 10%
MOL - Maximum normal operating load including design pressure, design temperature plus piping and support reactions.
PIT = Normal planned thermal transients associated with expected Plant normal operating transients such as start~up,, shutdown and load swings.
- MIT =_Maximum thermal transients in the systems functioning during Plant emergency conditions suchas full power reactor trip, turbine generator trip, loss of auxiliary power and the DBA.
- Design seismic load resulting fro_m a seismic ground surface acceleration of 0.1 g.
2SL = Hypothetical seismic load resulting from as seismic ground surface acceleration of 0.2 g.
The maximum stress* level of 10% over the miriimum yield stress at*
temperature for load combination No 3 is used to limit the strain to a value corresponding to the stress level reached by extrapolation
- - *beyon*d* *the material* proportioned *limit in accordance with-applicable codes.
Palisades FSAR DRAFT Proposed Changes Being Submitted With the Response to IE Bulletin 79-14 A-5
Palisades IEB 79-14
. The containment penetration assemblies are designed to accommodate the forces and moments due to pipe rupture.
Guides, pipe stops, increased.
pipe thickness or other means are provided to make the penetration the strongest part of the system.
The structural criteria for the reactor internals unde.r combined DBA and hypothetical seismic. loads are as follows:
- 1.
The core remains in place under the hydraulic loadings resulting from a complete break in one of the main coolant lines.
- 2.
Capability of the safety injection system to supply cooling water to the core is preserved.
- 3.
Certain distortions of the core stipport_structures occur but Conclusions 1 and 2 are still valid.
. The approach employed in assessing the capability of the core support structures to meet the requirements above is to determine the peak hydraulic loads during blowdown in a conservative manner and then to compare these peak dynamic loads with the static load-carrying capability of critical areas of the core suppbrt structures.
As the load is applied, it frequently occurs that two or more of the members share the load.
Members can fail by rupture which is postulated to occur when the allow_able stress Jevels are exceeded, or by instability.as in the case of buckling, for example.
For.these ultimate loading conditions, the allowable stresses used in tension, compression and bending are the ultimate strength of the material in
- tension, For bearing, shear and torsion, 150, 60 and 80 percent of the ultimate strength is used, respectively.
Using these values, a minimum safety factor of two will be maintained on all critical internal structures resulting from the combined loadings due to weight, DBA and hypothetical seismic loads.
(2)
Class I piping is designed to the load combinations and acceptance criteria shown below.
The piping is designed to the USAS B31.l.O *
(1967) Power Piping Code with three modifications.
Firstly, ~he primary pipe stress.es incorp_orate a 0.75i factor. *This factor was
- introduced into the ANSI B31.l Power Piping Code. (1973) and into the* -
ASME Code,Section III, Subsection NC (1974) and is presently employed in both codes.
Secondly, the criteria includes a faulted allowable of 2.4Sh.. The 2.4Sh allowable.was introduced into the ASME Code,
- section III, Subsection NC with the 1976 winter addenda and is
_____________ pr~sently employed inthis code.
Thirdly, the allowable 1. lSY. is
. addres-sedinA~2-:b above:c Palisades FSAR DRAFT Proposed Changes Being Submitted With th.e Response to IE Bulletin 79-14 A-6
Palisades IEB 79-14 Service Allowable Stress DP sh P.+ DW sh TE SA p +
p +
p +
DP TE p
DW SL DW + SL l.2Sh DW + 2SL Greater of
- 1. lS or 2.4Sh y
DW + F l.2Sh
=
=
=
=
=
Design pressure hoop stress Thermal expansion stress Longitudinal pressure stress Deadweight stress 2SL =
Design seismic stress loading resulting from a seismic ground surface acceleration of 0.1 g.
Hypothetical seismic stress loading, resulting from a seismic ground surface acceleration of 0.2 g.
Sh = Allowable code stress at temperature S = Minimum yield strength at temperature y
SA = Allowable code stress range F = Stresses resulting from thrust force of main
- steam relief valves or pressurizer relief valves (3)
Class I pipe supports are designed to the USAS B3.1.1.0 (1967) code and
.the AISC Structural Steel Specification, Seventh Edition, 1970, for the SL and 2SL cases as follows:
- 1.
SL Case Component Structural Catalog
- 2.
2SL Case
__
- _. *---- __, _____ __.C_omponent Structural
.catalog.
Load Combination Greater of DW + TE +
SL or DW + SL Greater of DW + TE +
SL or DW + SL Load Combination Greater of DW + TE +
2SL or DW + 2SL Greater of OW + TE +
2SL or ow*+ 2SL Allowable Stress
- 1. lS 0.2S u Allowable Stress Small of 1. lS y or 1. 65S Palisades.FSAR DRAFT Proposed Changes Being Submitted With tl:i.e Response to IE Bulletin 79-14 A-6a
... -~.*
- l. -*
e Palisades IEB 79-14 DW = Load due to pipe deadweight TE-= Load due to pipe thermal expansion SL = design seismic loading resulting from a seismic ground surface acceleration of 0.1 g 2SL = Hypothetical seismic loading, resulting from a seismic ground surface acceleration of 0.2 g S = AISC
- allowable s.tress S = Minimum ultimate stress u
S = Minimum yield stress y
(c) - Wind and Earthquake Loads for Class I Structures, Systems and Equipment (1)
Wind Force (W)
The wind loads are determined from the fastest mile of wind for*a 100--
year occurrence as shown in Figure l(b) of Reference 2.. This is 90 mph at the Palisades site.
The Class I structur'es are designed,
.. however, to withstand a sustained wind velocity of 100 mph.
In addition, Class I structures (except the enclosure over the fuel storage facilities) are designed to resist the effects of a 300 mph tornado as follows:
Loadings due to a tornado to be used in the design of tornado-res istant shelters are as.follows, the load to be applied simultaneously:
- a.
Differential pressure between inside and outside of enclosed areas *- 3 psid (b~rst1ng).
~~------- ------- ----~----
-~-
PalisadesFSAR DRAFT Proposed Changes Being Submitted With the Response to IE Bulletin 79-14 A*6b